ML20064G517
| ML20064G517 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  (DPR-59-A-043, DPR-59-A-43) |
| Issue date: | 11/22/1978 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20064G520 | List: |
| References | |
| NUDOCS 7812110320 | |
| Download: ML20064G517 (45) | |
Text
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UNITED STATES y
- g NUCLEAR REGULATORY COMMISSION q
, j WASHINGTON, D. c. 20005
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4, '.....s POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 43 License No. DPR-59 C
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l.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Power Authority of the State of New York (the licensee) dated August 18, 1978 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and
.i E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-59 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 43, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
g 3.
This amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 047:n.. $h Thomas As Ippolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: November 22, 1978 m
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ATTACHMENT TO LICENSE AMENDMENT NO. 43 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert 11 11 vii vii 6
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10 10 12 12 13 13 17 17 18 18 19 19 20 20 23 23 27 27 28 28 29 29 30 30 31 31 35 35 39 39 41 41, 41 a 42 42
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43 43, 43a 46 46 47 47 58 58 72 72 72 73 94 94 102 102 103 103 123 123 124 124 130 130 135c 135D 143 143 152 152 245 245
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JAFNPP TABLE OF CONTENTS (cont'd)
I. age F.
Minimum Emergency Core Cooling System F.
122 Availability G.
Maintenance of Filled Discharge Pipe G.
122 H.
Average Planar Linear Heat Generation 123 Race (APLEGR)
H.
I.
Linear Heat Generation Rate (LHGR)
I.
124 J.
Thermal Hydraulic Stability J.
124a SURVEILLANCE LIMITING CONDITIONS FOR OPERATION _S REQUIRDiEhTS 3.6 Reactor Coolant System 4.6 A.
Thermal Limitations A.
136 B.
Pressurization Temperature B.
137 g-C.
Coolant Chemistry C.
139 D.
Coolant Leakage D.
141 i
E.
Safety and Safety / Relief Valves E.
142a F.
Structural Integrity F.
144 G.
Jet Pumps G.
144 H.
Jet Pump Flow Mismatch H.
145 I.
Shock Suppressors (Snubbers) 1.
145a 3.7 Containment Systems 4.7 165 A.
165 B.
Standby Gas Treatment System B.
181 C.
184 D.
Primary Containment Isolation Valves D.
185 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems 4.9 215 A.
Normal and Reserve A-C Power Systems A.
215 l
B.
ame'gency A-C Power System B.
216 l
C.
Diesel Fuel C.
218 D.
Diesel Generator Operability D.
220 E.
Station Batteries E.
221 F.
LPCI HOV Independent Power Supplies F.
222a 3.10 Core Alterations 4.10 227 A.
Refueling Interlocks A.
227 B.
Core Monitoring B.
230 C.
Spent Fuel Storage Pool Water Level C.
231 D.
Control Rod and Control Rod Drive Maintenance D.
231 3.11 Additional Safety Related Plant Capabilities 4.11 237 A.
Main Control Room Ventilation A.
237 B.
Crescent Area Ventilation B.
239 l
C.
Battery Room Ventilation C.
239 11 f
Amendment No. p 43 l
1
JAFNPP LIST OF FIGURES Figure Title Page 1.1-1 APRM Flow Bias Scram Relationship to Normal Operating 23 i
Conditions 3.1-1 Manual Flow Control 47a j
4.1-1 Graphical Aid in the Selection of an Adequate Interval 48 Between Tests 4.2-1 Test Interval vs. Probability of System Unavailability 87 3.4-1 Sodium Pentaborate Solution Volume-Concentration 110 Requirements 3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111
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3.5-1 MAPIRGR Versus Planar Average Exposure 134 Initial Core, Type 2 3.5-2 MAPLEGR Versus Planar Average Exposure 135 Initial Core, Type 3 3.5-3 MAPLHGR Versus Planar Average Exposure 135a Reload 1, 8D274L 3.5-4 MAPLEGR Versus Planar Average Exposure 135b Reload 1, 8D274H 3.5-5 MAPLEGR Versus Planar Average Exposure 135c Reload 2, 8DRB265L 3.5-6 MAPLEGR Versus Planar Average Exposure 135d Reload 2, 8DRB283 3.6-1 Reactor Vessel Thermal Pressurization Limitations 163 4.6-1 Chloride Stress' Corrosion Test Results at 500 F 164 6.1-1 Management organization Chart 259 6.2-1 Plant Staff Organization 260 Amendment No. h, 2[, 43
JAFNPP surveillance tests, checks, calibrations, and exam-V.
Electrically Disarmed Control Rod instions shall be performed within the specified
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surveillance intervals. These intervals may be ad-To disarm a rod drive electrically, the four justed i 25 percent. The operating cycle interval amphenol type plus connectors are removed J
as pertaining to instrument and electrical surveil-from the drive insert and witiidrawat solenoids lance shall never exceed 15 months. In cases where rendering the rod incapable of withdrawal.
the elapsed interval has exceeded 100 percent of the This procedure is equivalent to valving out the specified interval, the next surveillance interval drive and is preferred. Electrical disarming shall commence at the end of the original specified does not eliminate position indication, inte rval.
W.
High Pressure Water Fire Protection System U.
Thermal Paraseters The High Pressure Water Fire Protection System j
1.
Minimum critical power ratio (MCPR)-Ratio of consists of: a water source and pumps; and that power in a fuel assembly which is calcu-distribution system piping with associated post i
lated to cause some point in that fuel assembly indicator valves (isolation valves). Such to experience boiling transition to the actual valves include the yard hydrant curb valves and assembly operating power as calculated by ap-the first valve ahead of the water flow alarm p11 cation of the GEXL correlation (Reference device on each sprinkler or water spray subsystem.
X.
Staggered Test Basis 2.
Fraction of Limiting Power Density - The ratio of the linear heat generation rate (LHGR) ex-A Staggered Test Basis shall consist of:
ist'ng at a given location to the design LHCR for that bundle type. Design LHGR's are 18.5 a.
A test schedule for a systems, subsystems, KW/ft for 7x7 bundles and 13.4 KW/ft for 8x8 trains or other designated components ob-and 8x8R bundles, tained by dividing the specified test in-terval into n equal subintervals.
3.
Maximum Fraction of Limiting Power Density -
The Maximum Fraction of Limiting Power Density b.
The testing of one system, subsystem, train (HFLPD) is the highest value existing in the or other designated component at the be-core of the Fraction of Limiting Power Density ginning of each subinterval.
(FLPD).
4 '. Transition Boiling - Transition boiling means the boiling region between nucleate and film boiling. Transition boiling is the region in which both nucleate and film boiling occur in-termittently with neither type being completely stable.
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h, 36, gl. 43 Amendment No.
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1.1 SAFETY LIMITS 2.1 LIMITING SAFETY SYSTIM SEITINGS 1.1 FUEL CLADb1NG INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability:
Applicability:
The Safety Limits established to preserve The Limiting Safety System Settings ' apply the fuel cladding integrity apply to those to trip settings of the instruments and variables which monitor the fuel thennat devices which are provided to prevent the
- behavior, fuel cladding integrity Safety Limits from being exceeded.
Objective:
Objective:
The objective of the Safety Limits is to The objective of the Limiting Safety System establish limits below which the integrity Settings is to define the level of the process of the fuel cladding is preserved, variables at which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.
Specifications Specifications A.
Reactor Pressure > 785 psig and Core Flow A.
Trip Settings
> 107. of Rated The existence of a minimum critical power The limiting safety system trip settings l
ratio (MCPR) less than 1.07 shall constitute shall be as specified below:
violation of the fuel cladding integrity safety limit, hereafter called the Safety 1.
Neutron Flux Trip Settings Limit.
a.
IRM - The IRH flux scram setting shall be set at s 120/125 of full scale.
l AmendmentNo.fe, [. f, 43
x JAFNPP 1
1.1 (cont'd) 2.1 (cont'd)
A.1.b.
APRM Flux Scram Trip Setting (Refuel or Start & Hot Standby Mode)
'lhe APRM flux scram setting shall be s 15 percent of rated neutron flux, with the Reactor Mode Switch in Startup/ Hot Standby or Refuel.
D.
Core Thermal Power Limit (Reactor Pressure c.
APRM Flux Scram Trip Settings (Run Mode) s 785 psig)
(1)
Flow Referenced Neutron Flux Scram Trip when the reactor pressure is s 785 psig or Setting core flow is less than lot of rated, the core thermal power shall not exceed 25 When the Mode Switch is in the RUN positjon, percent of rated thermal power.
the APRM flow referenced flux scram trip setting shall be C.
Power Transient S < 0.66 W + 54%
To ensure that the safety Limit establisb*.d in Specification 1.1.A and 1.1.B is not where exceeded, each required scram shall be initiated by its expected scram signal.
S = Setting in percent of rated The Safety Limit shall be assumed to be thermal power (2436 MWt) exceeded when scram is accomplished by a means other than the expected scram signal.
W = Icop recirculation flow rate in percent of rated (rated loop re-circulation flow rate equals 34.2 x 106 lb/hr)
For no combination of loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 117% of rated thermal power.
8 Amendment No.
, 43
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i JAFNPP 1.1 (cont'd) 2.1 (cont'd)
D.
Reactor Water Level (Hot or Cold In the event of operution with a maximum Shutdown Condition) fraction of limiting power de,sity (MFLPD) greater than the fraction of sted power Whenaver the reactor is in the shutdown (FRP), the setting shall be dified as condition with irradiated fuel in the follows:
reactor vessel, the water level shall
- not be less than that corresponding to SS (0.66 W + 544)
FRP 18 in.
(-146.5 in. indicated level)
MFLPD above the top of the active fuel whea it is seated in the core, where:
FRP = fraction of rated thermal power (2436 MWt)
MFLPD = maximum fraction of limiting power density where the limiting power
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density is 18.5 KW/ft for 7x7 fuel and 13.4 KW/f t for 8x8 and 8x8R fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
(2) Fixed High Neutron Flux Scram Trip Setting When the Mode Switch is in the RUN position, the APRM fixed high flux scram trip setting shall be S $ 120% Power pd,3)d,43 Amendment No.
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JAFNPP i
O 1.1 (cont'd) 2.1 (cont'd)
A.I.d.
APRM Rod Block Trip Setting h e APRM Rod block trip setting shall be:
S $ 0.66 W + 42%
4 where S = Rod block setting in percent of thermal power (2436 MWt)
J W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals (34.2 x 106 lb/hr) )
In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
4 S S (0.66 W + 42%)
i FRP = fraction of rated thermal power (2436 MWt)
MFLPD = maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7x7 fuel and 13.4 KW/ft for 8x8 and 8x8R fuel.
%e ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is
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less than the design value of 1.0, in which t
case the actual operating value will be used.
i AmendmentNo.1/,M,43 10
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JAFNPP r1 1.1 BASES A. Reactor Pressure > 785 psia and Core Flov >
2 107. o f Ra t ed 1.1 FUEL CIADDING INTEGRITY
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l Onset of transition boiling results in a de-The fuel cladding integrity limit is set such crease in heat transfer from the clad and, that no calculated fuel damage would occur as therefore, elevated clad temperature and the i
a result of an abnomal operational transient.
possibility of clad failure. However, the Because fuel damage is not directly observ-existence of critical power, or boiling trans-t l.
able, a step-back approach is used to establish ition, is not a directly observable parameter a Safety Limit such that the minimum critical in an operating reactor. Therefore, the mar-power ratio (MCPR) is no less than 1.07.
MCPR >
gin to boiling transition is calculated from i
1.07 represents a conservative margin relative plant operating parameters such as core power, j
to the conditions required to maintain fuel core flow, feedwater temperature, and core cladding integrity. The fuel cladding is one power distribution. The margin for each fuel 1
of the physical barriers which separate radio-assembly is characterized by the critical power i
active materials from the environs. The in-ratio (CPR) which is the ratio of the bundle l
tegrity of this cladding barrier is related to power which would produce onset of transition I
its relative freedosi from perforations or boiling divided by the actual bundle power.
I cracking. Although some corrosion or use re-The minintan value of this ratio for any bundle j
lated cracking may occur during the life of in the core is the minimum critical power ratio i
the cladding, fission product migration from (HCPR). It is assioned that the plant operation this source is incrementally cianulative and is controlled to,the nominal protective set-continuously measurable. Fuel cladding, per-points via the instrissented variables, i.e.,
forations, however, can result from thermal normal plant operation presented on Figure i
l stresses which occur from reactor operation 1.1-1 by the nominal expected flow control significantly above design conditions and the line. The Safety Limit (MCPR of 1.07) has protection system safety settings. While sufficient conservatism to assure that in the fission product migration from cladding per-event of an abnormal operational transient l
foration is just as measurable as that from initiated from the MCPR operating limits speci-use related cracking, the thermally caused fled for the normal operating conditions in speci-cladding perforations signal a threshold, be-fication 3.1.5, more than 99.97, of the fuel rods in yond which still greater thermal stresses may the core are expected to avoid boiling transi-1 cause gross rather than incremental cladding tion. The margin between MCPR of 1.0 (onset of I
deterioration. Therefore, the fuel cladding transition boiling) and the Safety Limit is Safety Limit is defined with margin to the derived from a detailed statistical analysis j
conditions which would produce onset of trans-considering all of the uncertainties in i
itic Miling, (MCPR of 1.0).
These conditions monitoring the core operating state including l
rrr et a significant departure from the uncertainty in the boiling transition correlation ecMh.on intended by design for planned as described in Reference 1.
The uncertainties operation.
employed in deriving the Safety Limit are i
12 h,1[,N,f.43 f
Amendment No.
i
i 1.1 CASE 3 (cont'd) e JAFNPP
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i provided at the beginning of each fuel cycle.
B. Core Thermal Power Limit (Reactor Pressure I
Because the boiling transition correlation
< 785 psig) is based on a large quantity of full scale
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data there is a very high confidence that At pressures below 785 psig the co a elevation i
operation of a fuel assembly at the Safety pressure drop (0 power, O flow).s greater 4
Limit would not produce boiling transition.
than 4.56 psi. At low powers and flows this Thus, although it is not required to establish pressure differential is maintained in the j
the safety limit, additional margin exists bypass region of the core. Since the pres-between the Safety Limit and the actual sure drop in the bypass region is essentially i
occurrence of loss of cladding integrity.
all elevation head, the core pressure drop at low powers c.d flows will always be greater However, if boiling transition were to occur, than 4.56 psi. 3 Analyses show that with a clad perforation would not be expected. Cladding flow of 28 x 10 lbs/hr bundle flow, bundle temperatures would increase to approximately pressure drop is nearly independent of bundle l
11000F which is below the perforation temper-power and has a value of 3.5 psi. Thus, the ature of the cladding material. This has been bundle flow with a 4.56 psi driving head will
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verified by tests in the General Electric Test be greater than 28 x 103 lbs/hr. Full scale Reactor (CETR) where fuel similar in design ATLAS test data taken at pressures from 0 i
to Fitspatrick operated above the critical heat psig to 785 psig indicate that the fuel as-l flux for a significant period of time (30 min-sembly critical power at this flow is rpprox-utes) without clad perforation.
imately 3.35 MWt. With the design peaking factors this corresponds to a core thermal
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If reactor pressure should ever exceed 1400 psia power of more than 50%. Thus, a core thermal during normal power operating (the limit of power Ibnit of 25% for reactor pressures j
applicability of the boiling transition corre-below 785 psig is conservative.
lation) it would be assumed that the fuel cladding i
integrity Safety Limit has been violated.
In addition to the boiling transition limit (Safety Limit) operation is constrained to a i
maxiansa IJICR = 18.5 kw/ft for 7x7 fuel and l
13.4 kw/ft for 8x8 and 8x8R fuel. At 100%
j power, this Ibnit is reached with a maximum fraction of limiting power density (MFLPD)
)
equal to 1.0.
In the event of operation with i
a MFLPD greater than the fraction of rated power (FRP), the APRM scram and rod block settings shall be adjusted as required in Specifications 2.1.A.1.c and 2.1.A.1.d.
I 13
[, M. [, 43 Anenament No.
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JAFNPP j
2.1 BASES (cont'd)
In order to ensure that the IRM provided adequate pro-high local peaks, and because several rods must be tection against the single rod withdrawal error, a moved to change power by a significant percentage 1
range of rod withdrawal accidents was analyzed. This of rated power, the rate of power rise is very analysis included starting the accident at various power slow. Cencrally, the heat flux is in near equili-levels, n e most severe case involves an initial con-brium with the fission rate.
In an asstaned uniform dition in which the reactor is just suberitical and rod withdrawal approach to the scram level, the i
the IRM system is not yet on scale. This condition rate of power rise is no more than 5 percent of j
exists at quarter rod density. Additional conservatism rated power per minute, and the APRM system would j
was taken in this analysis by asstaning that the IRM be more than adequate to assure a scram before j
channel closest to the withdrawn rod is by-passed.
the power could exceed the safety limit. The 15 l
The results of this analysis show that the reactor is percent APRM scram remains active until the mode l
scrammed and peak power limited to one percent of switch is placed in the RUN position. Thin switch
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l rated power, thus maintaining MCPR above the Safety occurs when reactor pressure is greater than 850 psig.
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I 1.imit.
Based on the above analysis, the IRM provides protection against local control rod withdrawal errors
- c. APRM Flux Scram Trip Setting (Run Mode) and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
The APRM flux scrara trip in the run mode consists j
of a flow referenced scrma setpoint and a fixed b.
APRM Flux Scram Trip Setting (Refuel or Startup and.
high neutron flux scran setpoint. The APRM flow Hot Standby Mode) referenced neutron flux signal is passed through a filtering network with a time constant which is l
For operation in the startup mode while the reactor is representative of the fuel dynanics. This provides at low pressure, the APRM scram setting of 15 percent a flow referenced signal that approximates the of rated power provides adequate themal margin between average heat flux or thermal power that is developed the setpoint and the safety limit, 25 percent of rated.
in the core during transient or steady-state condi-The margin is adequate to accomanodate anticipated tions. This prevents spurious scrams, which have maneuvers associated with power plant startup. Effects an adverse effect on reactor safety because of the of increasing pressure at zero or low void content are resulting themal stresses. Examples of events minor, cold water from sources available during startup which can result in momentary neutron flux spikes is not much colder than that already in the system, are momentary flow changes in the recirculation temperature coefficients are small, and control rod system flow, and small pressure disturbances patterns are constrained to be uniform by operating during turbine stop valve and turbine control procedures backed up by the rod worth minimizer and the valve testing. These flux spikes represent no Rod Sequence Control System. Worth of individual rods hazard to the fuel since they are only of a few i
is very low in a uniform rod pattern. Thus, of all seconds duration and less than 120% of rated themal possible sources of reactivity input, uniform control pwer.
rod w".thdrawal is the most probable cause of signifi-l cant power rise. Because the flux distribution asso-The APRM flow referenced scran trip setting at full j
cla:ed with unifom rod withdrawals does not involve recirculation flow is adjustable up to 117% of AmendmentNo.[,
, [, 43 i
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2.1 BASES (ccat'd)
JAFNPP i
- d. APRM Rod Block Trip Setting i
4 J
rated power. This reduced flow referenced trip setpoint Reactor power level may be varied by moving control will result in an earlier scrani during slow thennal rods or by varying the recirculation flow rate. The transients, such as the loss of 80 F feedwater heating APRM system provides a control rod block to prevent
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event, than would result with the 1207. fixed high rod withdrawal beyond a given point at constant re-I j
neutron flux scraan trip. H e lower flow referenced g
circulation flow rate, and thus to protect against 1
scrant setpoint therefore decreases the severity (ACPR) l the condition of a MCPR less than the Safety Limit, of a slow thermal transient and allows lower Operating This rod block trip setting, which is automatically Limits if such a transient is the limiting abnormal varied with recirculation loop flow rate, prevents operational transient during a certain exposure in-an increase in the reactor power level to excessive i
terval in the cycle.
values due to control rod withdrawal. He flow variable trip setting provides substantial margin l
% e APRM fixed high neutron flux signal does not in-from fuel damage, assuming a steady-state operation corporate the time constant, but responds directly to at the trip setting, over the entire recirculation instantaneous neutron flux, h is scram setpoint scrams flow range. The margin to the Safety Limit in-the reactor during fast power increase transients if creases as the flow decreases for the specified credit is not taken for a direct (position) scran, and trip setting versus flow relationships therefore also serves to scram the reactor if credit is not taken the worst case MCPR which could occur during steady-for the flow referenced scram.
state operation is at 1087. of rated thermal power because of the APRM rod block trip setting. The j
The scrasi trip setting must be adjusted to ensure that actual power. distribution in the core is established the LHGR transient peak is not increased for any com-by specified control rod sequences and is monitored bination of maxistan fraction of limiting power density continuously by the in-core LPRM system. As with i
(MFLPD) and reactor core thermal power. The scram set-the APRM scram trip setting, the APRM rod block
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ting is adjusted in accordance with the formula in trip setting is adjusted downward if the maximssa i
Specification 2.1.A.1.c, when the MFLPD is greater than fraction of limiting power density exceeds the frac-l the fraction of rated power (FRP). This adjustment tion of rated power, thus preserving the APRM rod
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may be accomplished by either (1) reducing the APRM block safety margin. As with the scrma setting, I
scran and rod block settings or (2) adjusting the in-this may be accomplished by adjusting the APRM i
dicated APRM signal to reflect the high peaking
- gain, condition.
l
- 2. Reactor Water Low Level Scran Trip Settina (LLI) j Analyses of the limiting transients show that no scram adjustment is required to assure that the MCPR will be The reactor low water level scran is set at a point greater than the Safety Limit when the transient is which will assure that the water level used in the i
initiated frosi the MCPR operating limits provided in Bases for the Safety Limit is maintained. The Specification 3.1.B.
scran setpoint is based on normal operating temp-erature and pressure conditions because the level instrumentation is density compensated.
18 4
Amendment No f. f.
, 43 i
I l
2.1 BASES (cont'd) e JAFNPP s
- 3. Turbine Stop Valve Closure Scr1s Trip Settings
- 5. Main Steam Line Isolation valve closure Scram Trip l
Setting
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The turbine stop valve closure scram trip anticipates j
the pressure, neutron flux and heat flux increase that The low pressure isolation of the main steam lines could result from rapid closure of the turbine stop at 825 psig was provided to give protection against l
valves. With a scram trip setting of s 10 percent of rapid reactor depressurization and the resulting i
velve closure from full open, the resultant increase in rapid cooldown of the vessel. Advantage was taken of surface heat flux is limited such that MCPR remains the scram feature which occurs when the main steam I
l along the Safety Limit even during the worst case line isolation valves are closed, to provide for j
s transient that assumes the turbine bypass is closed.
reactor shutdown so that high power operation at low j
n is scram is bypassed when turbine steam flow is below reactor pressure does not occur, thus providing pro-l 30% of rated, as measured by turbine first stage tection for the fuel cladding integrity safety limit.
pressure.
Operation of the reactor at pressures lower than 825 j
psig requires that the Reactor Mode Switch be in the j
- 4. Turbine Control Valve Fast Closure Scram Trip Setting Startup position where protection of the fuel clad-ding integrity safety limit is provided by the APRM This turbine control valve fast closure scram antici-high neutron flux scram and the IRM. Thus, the com-pates the pressure, neutron flux, and heat flux in-bination of main steam line low pressure isolation crease that could result from fast closure of the tur-and isolation valve closure scram assures the avail-l bine control valves due to load rejection exceeding the ability of neutron flux scram protection over the capability of the turbine bypass. H e Reactor Protec-entire range of applicability of the fuel cladding j
tion System initiates a scram when fast closure of the integrity safety. limit. In addition, the isolation control valves is initiated by the. fast. acting solenoid valve closure scram anticipates the pressure and valves. This is achieved by the action of the fast flux transients which occur during normal or inad-
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acting solenoid valves in rapidly reducing hydraulic vertent isolation valve closure. With the scrams j
control oil pressure at the main turbine control set at s 10 percent valve closure, there is no in-valve actuator dise dsasp valves. This loss of pres-crease in neutron flux.
I sure is sensed by pressure switches whose contacts
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form the one-out-of-two-twice logic input to the re-
- 6. Main Steam Line Isolation Valve Closure on Low Pressure j
actor protection system. This trip setting, a j
nominally 50 percent greater closure time and a dif-The low pressure isolation minimum limit at 825 psig i
ferent valve characteristic from that of the turbine was provided to give protection against fast reactor l
stop valve, combine to produce transients very similar depressurization and the resulting rapid cooldown of and no more severe than for the stop valve. No signifi-the vessel. Advantage was taken of the scram feature l
cant change in MCPR occurs. Relevant transient which occurs when the main steam line isolation valves analyses are discussed in Section 14.5 of the Final are closed to provide for reactor shutdown so that j
Safety Analysis Report. This scram is bypassed when Operation at pressures lower than those specified in turbine steam flow is below 30 percent of rated, as the thermal hydraulic safety limit does not occur, measured by turbine first stage pressure.
although operation at a pressure lower than 825 psig would not necessarily constitute an unsafe condition, i
19 Amendment No. 1[e, f, h, f, 43 1
i i
2.1 CASES (cont'd)
JAFNPP C. References J
1.
Linford, R. B., "An 'ftical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor, "NEDO-10802, Feb., 1973.
2.
Licensing Topical Reports, " General Electric Boil-ing Water Reactor Generic Reload Fuel Application".
NEDO-24011-2, March, 1978.
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20 0
d.3,43 Amendment No.
RATED THERMAL PokTR = 2436 RATED CORE FLOW
= 77.0 x 10 120 p -
.- =
BIAS SCRAM g
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100 y
g
/
i d
p EE
/
h
/
se 80 g
l NOMINAL EXPECTED l
l FLOW CONTROL LINE
)
8
/
h 60 l
(
y
=
8 s
CORE 40 NEUTRON
~ PokTR LIMIT 2S%
1 20 NATURAL CIRCULATION LINE A
g 0
20 40 60 80 100 120 CORE FIDW RATE
(% OF RATED)
FIGURE 1.1-1 APRM FLOW BIAS SCRAM RELATIONSHIP 2D NORMAL OPERATING CONDITIONS Amendment No.
43 23 (next page is 27)
l JAFNPP 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR C00lJuff SYSTEM APPLICABILITY:
APPLICABILITY:
l Applies to limits on reactor coolant Applies to trip settings of the instru-system pressure.
ments and devices which are provided to prevent the reactor coolant system safety i
limits from being exceeded.
OBJECTIVE:
OBJECTIVE:
j To establish a limit below which the To define the level of the process integrity of the Reactor Coolant System variables at which automatic protective is not threatened due to an overpressure action is initiated to prevent the safety condition.
limits from being exceeded, j
SPECIFICATION:
SPECIFICATION:
j
}
1.
The reactor coolant system pressure 1.
The Limiting Safety System setting j
shall not exceed 1,325 psig at any shall be specified below:
1 time when irradiated fuel is present 4
in the reactor vessel.
A.
Reactor coolant high pressure scram shall be $ 1,045 psig.
B.
Reactor coolant system safety / relief valve nominal settings shall be as follows:
i Safety / Relief Valves 2 valves at 1090 psig 2 valves at 1105 usig j
7 valves at 1140 psia l
The allowahle setpoint error for each safety / relief valve shall be + 1 percent.
27 l
Amendment No. V6, )$. 43 I
1 c '
JAFNPP i
1.2 (cont'd) 2.2 (cont'd) 2.
We reactor vessel done pressure shall not 2.
Action shall be taken to decrease the exceed 75 psig at any time when operating reactor vessel dome pressure below 75 the Residual Heat Removal pump in the psig or the shutdown cooling isolation shutdown cooling mode.
valves shall be closed.
I 3
4 I
I 1
4 i
1
[, 43 i
Amendment No.
28
JAFNPP 1.2 and 2.2 BASES The reactor coolant pressure boundary ANSI Code permits pressure transients up to integrity is an important barrier in 20 percent over the design pressure (1204 x the prevention of uncontrolled release 1,150 = 1,380 psig). The safety limit i
of fission products. It is essential pressure of 1,375 psig is referenced to the lowest elevation of the Reactor Coolant System.
that the integrity of this boundary be protected by establishing a pres-The analysis in NEDO-24129, " Supplemental Reload sure limit to be observed for all Licensing Submittal for the James A. FitzPatrick operating conditions and whenever Nuclear Power Plant for Reload No.
2", June 1978, there is irradiated fuel in the as amended by NEDO-24129-1. Supplement 1, September I
reactor vesret.
1978, shows that the main steam isolation valve l
The pressure safety limit of 1,325 psig transient, when direct scram is ignored, is the most severe event resulting directly in a reactor as measured by the vessel steam space pressure indicator is equivalent to coolant system pressure increase. The reactor 1,375 psig at the lowest elevation of vessel pressure code limit of 1,375 psig, given in FSAR Section 4.2, is above the peak pressure the Reactor Coolant System. The 1,375 psig value is derived from the produced by the event above. Thus, the pressure design pressures of the reactor pres-safety limit (1,375 psig) is well above the peak pressure that can result from reasonably expected sure vessel and reactor coolant system piping. The respective design pressures l
overpressure transients. Figure 3 in NEDO-24129-1 presents the curve produced by this analysis.
0 are 1250 psig at 575 r for the reactor Reactor pressure is continuously indicated in 0
vessel, 1148 psig at 568 F for the recirculation auction piping and 1274 the control room during operation.
psig at 5750F for the discharge piping.
The pressure safety limit was chosen A safety limit is applied to the Residual Heat as the lower of the pressure tran-Removal system (RHRS) when it is operating in the shutdown cooling mode. When operating in sients permitted by the applicable the shutdown cooling mode, the RffRS is included design codes: 1965 ASME Boller and Pressure vessel Code,Section III for in the reactor coolant system.
the pressure vessel and 1969 ANSI B31.1 Code for the reactor coolant system pip-ing. The ASME Boiler and Pressure Vessel Code permits pressure transients up to 10 percent over design pressure (110s x 1,250 = 1,375 psig), and the Amendment No. 15, 25, 37, 43 29
rs i
JAFNPP I
3.1 f.IHITING CONDITIONS FOR OPERATION 4.1 SURVEII. LANCE REQUIRENENTS
~
- 3. 's REACTOR PRUTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTDI 4
l Applicability:
Applicability:
Applies to the instrumentation and associated Applies to the surveillance of the instru-devices which initiate the reactor scram.
mentation and associated devices which initiate reactor scram.
Objective:
i Oblective:
To assure the operability of the Reactor j
Protection System.
To specify the type of frequency of i
surveillance to be applied to the protection
}
Specification:
instrumentation.
A.
The setpoints, minimum number of trip Specification:
systems, ministan nisnber of instrisment channels that must be operable for each A.
Instrumentation systems shall be position of the reactor mode switch shall be functionally tested and calibrated as as shown on Table 3.1-1.
The design system indicated in Tables 4.1-1 and 4.1-2 response time from the opening of the sensor respectively, contact to and including the opening of the trip actuator contacts shall not exceed 100 j
meec.
j B.
Minimin Critical Power Ratio (HCPR)
B.
Maximum Fraction of Limiting Power i
Density (HFIED)
During reactor power operation at rated power and flow, the HCPR operating The NFLPD shall be determined daily during limits shall not be less than those shown below:
reactor power operation at 2: 25% rated thermal power and the APRM high flux scram FUEL HCPR OPERATING l.IMIT FOR INCRENENTAL and Hod Block trip settings adjusted if TYPE CYCI.E 3 CORE AVERACE EXPOSURE necessary as regulred by Specifications 2.1.A.I.c and 2.1.A.I.d, respectively.
BOC3 to 2GWd/t EOC3-2CWJ/t EOC3-1CWd/t before EOC3 to EOC3-lCWd/t to EOC3 7x7 1.21 1.25 1.30 8x8 1.22 1.33 1.37 8x8R 1.20 1.33 1.37 1/., @, [, %, M, 43 j
Amendaent No.
30
x JAFNPP 3.1 (Cont'd)
If anytime during reactor operation greater than 25% of rated power it is determined that the limiting value for HCPR is'being exceeded, action shall then be initiated within fifteen (15) minutes to restore operation to within the prescribed limits. If the MCPR is not returned to within the prescribed limits within two (2) hours, an orderly reactor power reduction shall be commenced immediately.
The reactor power shall be reduced to less than 25%
of rated power within the next four hours, or until the MCPR is returned to within the prescribed limits. For core flows other than rated, the HCPR operating limit shall be multiplied by the appro-priate kg factor where kg is as shown in figure 3.1.1.
C.
MCPR shall be determined daily during reactor power operation at
= 25% rated thermal power and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for Specification 3.3.B.5.
l D.
When it is detensined that a channel has failed in the unsafe condition, the other RPS channels that monitor the same variable shall be functionally tested immediately before the trip system containing the failure is tripped. The trip system containing the unsafe failure may be placed in the untripped condition during the period in which surveillance testing is being performed on the other RPS channels.
j4,gd,yi,p0,3,43 8
31 I
Amendment No.
[
~'
3.1 BASES (cont'd)
Turbine control valves fast closure initiates a scram based on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The switches are located between fast closure solenoids and the disc dump valves, and are set relative (500 < P < 850 psig) to the normal EHC oil pressure of 1,600 psig so that, based on the small system volume, they can rapidly detect valve closure or loss of hydraulic pressure.
The requirement that the IRH's be inserted in the core when the APRH's read 2.5 indicated on the scale in the startup and refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation.
B. The limiting. transient which determines the required steady state HCPR limit de-pends on cycle exposure. The operating limit NCPR values as detenmined from the 3
transient analysis for Cycle 3 (NEDO-24129 l
and NEDO-24129-1, Supplement 1) for various l
core exposures are given in Specification 3.1.B.
The ECCS performance analysis mastened reactor operation will be limited to MCPR o f 1.18.
However, the Technical Spect-fications limit operation of the reactor to the more conservative HCPR based on consideration of the limiting transient as given in Specification 3.1.B.
a send nent no. p., p, 2/, f, /, 43
(-
JAFNPP r'
4.1 BASES (cont'd)
Calibration on this frequency assures plant operation at or below is meaningful to the one performed thermal limits.
just prior to shutdown or startup; i.e.,
the tests that are performed A comparison of Tables 4.1-1 and just prior to use of the instrument.
4.1-2 indicates that tso instrument channels have not been Calibration frequency of the included in the latter table. These instrument channel is divided into are: mode switch in shutdown and two groups. These are as follows:
manual scram. All of the devices or sensors associated with these scram 1.
Passive tyre indicating devices functions are simple on-off switches that can be compared with like and, hence, calibration during units on a continuous basis.
operation is not applicable.
2.
Vacuum tube or semi-conductor B. The MFLPD is checked once per day to devices and detectors that determine if the APRM scram requires drift or lose sensitivity.
adjustment. Only a small number of control rods are moved daily and thus Experience with passive type the MFLPD is not expected to change instruments in generating stations significantly and thus a daily check and subitations indicates that the of the MFLPD is adequate.
specified calibrations are adequate.
For those devices which employ amplifiers, etc., drift specifica-tions call for drift to be less than 0.4 percent / month; i.e.,
in the period The sensitivity of LPN1 detectors of a month a m'aximum drift of 0.4 decreases with exposure to neutron percent could occur, thus providing flux at a slow and approximately for adequate margin, constant rate. This is compensated for in the APRM system by For the APRM System, drift of calibrating twice a week using heat electronic apparatus is not the only balance data and by calibrating consideration in determining a individual LPRM's every 1000 effec-calibration frequency. Change in tive full power hours, using TIP power distribution and loss of traverse data.
chamber sensitivity dictate a calibration every 7 days.
l Amendment No. 43
JAFWPP TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT
~
Minimum No, Modes in Which Total cf Operable Trip Level Function Must be Number of Instrument Trip Function Setting Operable Instruisent Action Channels Channels (1) per Trip Refuel Startup Run Provided System (1)
(6) by Design for Both Trip Systems 1
Mode Switch in X
X X
1 Mode Switch A
Shutdown (4 Sections) 1 Manual Scram X
X X
2 Instrument A
Channels 3
IRM High Flux s 120/125 of full scale X
X 8 Instrument A
Channels 3
IRM Inoperative X
X 8 Instrument A
Channels 2
APRM Neutron Flux-s 15% Power X
X 6 Instrument A
Startup(15)
Channels 2
APRM Flow Referenced Sn ".(0.66W+547.)x X
6 Instrument A or B Neutron Flux (12) (13)
FRP ~
Channels (14)(Not to exceed 117%)
MFLPD 2
APRM Fixed iilgh Neutron s 120% Power X
6 Instrument A or B Flux (14)
Channels 2
APRM Inoperative (10)
X X
X 6 Instrument A or B Channels Amendment No.1[, h, 43
~
JAFNPP rs 2 TABLE 3.1-1 (Cont'd)
REACTOR PROTECTION SYSTEN (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.
Modes in Which Total of Operable Function Must Be Number of Instrument Trip Level Operable Instrument Channels Trip Function Setting Channels Action per Trip Refuel Startup Run Provided (1)
System (1)
(6) by Design for Both Trip Systems 2
APRM Downscale 2 2.5 indicated on X
6 Instrument A or B scale (9)
Channels
~
2 High Reactor
$ 1045 psig X(8)
X X
4 Instrument A
Pressure Channels 2
High Drywell 5 2.7 psig X(7)
X(7)
X 4 Instrument A
Pressure Channels 2
Reactor Low Water 212.5 in.
X X
X 4 Instrument A
Level indicated level Channels 2
High Water Level 5 36 gal X(2)
X X
4 Instrument A
l in Scram Discharge Channels Volume 2
$ 3 x nomal full X X
X 4 Instrument A
High Radiation power background Channels 4
Main Steam Line 5 10% valve X(3)(5)
X(3) (5)
X(5) 8 Instrument A
Isolation Valve closure Channels Closure 2
Turbine Control 500< P < 850 psig X(4) 4 Instrument A or C Valve Fast Control oil pressure Channels Closure between fast closure solenoid and disc dump valve AmendmentNo.p,43
JAFNPP ABLE 3.1-1 (Cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.
Modes in Which Total of Operable Function Must be Ntunber of Instrument Trip Level Operable Instrument Channels Trip Function Setting Channels Action per Trip Refuel Startup Run Provided (1)
System (1)
(6) by Design for Both Trip Systems 4
Turbine Stop s 10 % valve X(4)(5) 8 Instrument A or C Valve closure closure Channels NOTES OF TABLE 3.1-1 1.
There shall be two operable or tripped trip systems for each function, except as specified in 4.1.D.
From and after the time that the minimum number of operable instrument channel for a trip system cannot be met,theaffected trip system shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.
A.
Initiate insertion of operable rods and complete insertion of all operable rods within four hours.
B.
Reduce power level to IRM range and place Mode Switch in the Startup Position within eight hours.
C.
Reduce power to less than 30 percent of rated.
2.
Permissible to bypass, in Refuel and Shutdown positions of the Reactor Mode Switch.
3.
By passed when reactor pressure is < 1005 pais.
4.
Bypassed when turbine first stage pressure is less than 217 psig or less than 30 percent of rated.
5.
The design permits closure of any two lines without a scram being initiated.
6.
When the reactor is subcritical and the reactor water temperature is less than 212 F, only the following trip functions need to be operable:
a A.
Mode Switch in Shutdown B.
Manual Scram Amendment No. 43 42
JAFNPP e'
JlLE 3.1-1 (Cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT NOTES OF TABLE 3.1-1 (Cont'd)
C.
High Flux IRM D.
Scram Discharge Volume High Level E.
APRM 15% Power Trip 7.
Not required to be operable when primary containment integrity is not required.
1 i
i 8.
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel l
9.
The APRM downscale trip is automatically bypassed when the IRM Instrumentation is operable and not high.
l 10.
An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 11 LPRM inputs of the normal complement.
11.
See Section 2.1.A.I.
12.
This equation will be used in the event of operation with a maximum frsction of limiting power density (NFLPD) greater than the fraction of rated power (FRP).
where:
Fraction of rated thermal power (2436 MWt)
=
MFLPD = Maximum fraction of limiting power density where the limiting power density is 18.5 KW/ft for 7x7 fuel and 13.4 KW/ft for 8x8 and 8x8R fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than I
the design value of 1.0, in which case the actual operating value will be used 1
W
= Loop Recirculation flow in percent of rated (rated is 34.2 x 106 lb/hr) a S,
Scram setting in percent of initial
=
13.
The Average Power Range Monitor scram function is varied (Figure 1.1-1) as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 4
2.1.A.1.c.
d 3[, 3, 43 Amendment No.
d
,- 3 TABLE 3.1-1 (Cont'd)
REACTOR PROTECTION SYSTEM _(SCRAM) INSTRUMENTATION REQUIREMENT l
NOTES OF TABLE 3.1-1 (Cont'd)
- 14. The APRM flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. The APRM fixed high neutron flux signal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
- 15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is placed in the Run position.
l l
l Amendment No. 43 43a
JAFNPP O
Table 4.1-2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INS 11tUHErn CHANNELS 1
Instrtament Channel Croup (1)
Calibration (4)
Minimum Frequency (2)
IRM High Flux C
Comparison to APRM on Maximum frequency once/ week Controlled Shutdowns APRM High Flux Output Signal B
Ileat Balance Daily Flow Bias Signal B
Internal Power and Every refueling outage Flow Test with Stan-dard Pressure Source LPRM Signal B
TIP Systesa Traverse Every 1000 effective full power hours High Reactor Pressure A
Standard Pressure Every 3 months Source High Drywell Pressure A
Standard Pressure Every 3 months Source Reactor Low Water Level A
Pressure Standard Every 3 months High Water Level in Scram Dis-A Note (5)
Note (5) charge Voltane Main Steam Line Isolation Valve A
Note (5)
Note (5)
Closure Main Steam Line High Radiation B
Standard Current Every 3 months Source (3)
Turbine Plant Stage Pressure A
Standard Pressure Every 6 months Permissive Source l
Turbine control Valve Fast Closure A
Standard Pressure Once/ operating cycle Oil Pressure Trip Source Amendment No. 43 46
JAFNFP n
Teble 4.1-2 (cent'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Group (1)
Calibration (4)
Minimtan Frequency (2)
Turbine Stop Valve Closure A
Note (5)
Note (5)
Reactor Pressure Permissive A
Standard Pressure Every 6 months Source NOTES FOR TABLE 4.1-2 1.
A description of three groups is included in the Bases of this Specification.
2.
Calibration test is not required on the part of the system that is not required to be operable, or is tripped, but is required prior to return to service.
3.
The current source provides an instrisnent channel alignment. Calibration using a radiation source shall be made each refueling outage.
4.
Response time is not a part of the routine instrtsnent channel test but will be checked once per operating cycle.
5.
Actuation of these switches by normal means will be perfomed during the refueling outages.
e i
Amendment No. 43 47
JAFNPP rx c,
(
)
3.2 BASES (cent'd)
~
crease to the Safety Limit. The trip logic The scaling arrangement is such that for this function is 1 out of n:
e.g., any trip setting is less than a factor of 10 trip on one of six APRM's, eight IRM's, or above the indicated level.
four SRM's will result in a rod block.
A downscale indication on an APRM or IRM The minimum instrument channel requirements is an indication the instrument has assure sufficient instrumentation to assure failed or the instrument is not the single failure criteria is met.
The sensitive enough. In either case the minimum instrument channel requirements for instrument will not respond to changes the RBM may be reduced by one for mainten-in control rod motion and thus, control ance, testing, or calibration. This time rod motion is prevented. The downscale period is only three percent of the oper-trips are set at 2.5 indicated on scale.
ating time in a month and does not signifi-cantly increase the risk of preventing an The flow comparator and scran discharge inadvertent control rod withdrawal.
volume high level components have only one logic channel and are not required The APRM rod block function is flow for safety. The flow comparator must be biased and prevents a significant bypassed when operating with one re-reduction in HCPR especially during circulation water pump.
operation at reduced flow. The APRM provides gross core protection; i.e.,
The refueling interlocks also operate limits the gross core power increase one logic channel, and are required for from withdrawal of control rods in the safety only when the Mode Switch is in normal withdrawal sequence. The trips the Refueling position.
are set so that MCPR is maintained greater than the Safety Limit.
For effective emergency core cooling for small pipe breaks, the HPCI system must The RBM rod block function provides function since reactor pressure does not local protection of the core:
1.e., the decrease rapidly enough to allow either prevention of boiling transition in a core spray or LPCI to operate in tbne.
local region of the core, for a Jingle The automatic pressure relief function rod withdrawal error from a limiting is provided as a backup to the HPCI in control rod pattern.
the event the HPCI does not operate.
The arrangement of the tripping contacts The IRM rod block function provides is such as to provide this function when local as well as gross core protection.
necessary and minimize spurious operation. The trip settings given in AmendmentNo.g6,yk,1,43 6
.JAFNPP o
)
ABLE 3.2-3 INSTRUNDITATION THAT INITIATES CONTROL ROD BLOCKS Minissa No.
Total Number of Instrument of Operable Channels Provided by Design Instrument Instrument Trip Level Setting for Both Channels Action Channels Per Trip System 2
APRM Upscale (Flow S s (0.66W+42%)x 6 Inst. Channels (1)
Blased)
-M PD-2 APRM Upscale (Start-s 127.
6 Inst. Channels (1) up Mode) 2 APPJi Downscale 2 2.5 indicated on scale 6 Inst. Channels (1) 1 (6)
Rod Block Monitor S s 0.66Wi-397.(8) 2 Inst. Channels (1)
(Flow Biased) 1 (6)
Rod Block Monitor 22.5 indicated on scale 2 Inst. Channels (1)
Downscale 3
IRM Downscale (2) 2 2% of full scale 8 Inst. Channels (1) 3 IRM Detector not in (7) 8 Inst. Channels (1)
Startup Position 3
IRM Upscale 5 06.4%'of full scale 8 Inst. Channels (1) 2 (4)
SRM Detector not in (3) 4 Inst. Channels (1)
Startup Position
- 2. (4) (5)
SRM Upscale s10 counts /sec 4 Inst. Channels (1)
NOTES FOR TABLE 3.2-3 1.
For the Startup and Run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function. The SRM and IRM blocks need not be operable in run mode, and f, f 43 Amendment No.
JAFNPP TABLE 3.2-3 (Cont'd)
INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3 (Cont'd) the APRM and RBM rod blocks need not be operable in startup mode. From and after the time it is found that the first column cannot be met for one of the two trip systems, this condition may
^
exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped. From and after the time it is found that the first column cannot be met for both trip systems, the systems shall be tripped.
l 2.
IRM downscale is bypassed when it is on its lowest range.
3.
This function is bypassed when the count rate is = 100 cps.
4.
One of the four SRM inputs may be bypassed.
5.
This SRM Function is bypassed when the IRM range switches are on range 8 or above.
i 6.
The trip is bypassed when the reactor power is s: 30%.
7.
This function is bypassed when the Mode Switch is placed in Run.
8.
S = Rod Block Monitor Setting in percent of initial 6
W = Loop recirculation flow in percent of rated (rated loop recirculation flow is 34.2 x 10 lb/hr).
Amendment No.
. 43 p
JAFNPP TN
)
3.3 (Cont'd) 4.
Control rods shall not be 4.
Prior to control rod withdrawal withdrawn for startup or for startup or during refueling, refueling unless at least verify that at least two source l
two source range channels range channels have an observed have an observed count count rate of at least three rate equal to or greater counts per second.
than three counts per second.
5.
During operation with 5.
When a limiting control rod limiting control rod pattern exists, an instrument patterns, as determined by functional test of the RBM shall the designated qualified be performed prior to withdrawal i
l personnel, either:
of the designated rod (s).
l a.
Both RBM channels shall be operable, or b.
Control rod withdrawal shall be blocked, or 1
c.
The operating power level shall be l
limited so that MCPR will remain above the Safety Limit assuming a single error that results in complete withdrawal of any single operable control rod.
. h, 43 Amendment No. 1
JAFNPP A
l 3.3 cnd 4.3 BASES (cent'd)
This system'bteks up the operctor who withdraws control rods according to written sequences.
rods have been withdrawn (e.g. groups A12 and The specified restrictions with one channet j
A34), it is demonstrated that the Group Notch out of service conservatively assure that fuel made for the control drives is enforced. This damage will not occur due to rod withdrawat demonstration is made by performing the hardware errors when this condition exists, functional test sequence. The Group Notch re-straints are automatically removed above 20% power.
A limiting control rod pattern is a pattern which results in the core being on a thermal During reactor shutdown, similar surveillance hydraulic limit (i.e. MCPR limits as shown in checks shall be made with regard to rod group specification 3.1.B).
During use of such availability as soon as automatic initiation of patterns, it is judged that testing of the the RSCS occurs and subsequently at appropriate RBH System prior to withdrawal of such rods to stages of the control rod insertion, assure its operability will assure that in-proper withdraw does not occur. It is the 4.
The Source Range Monitor (SRM) System performs responsibility of the Reactor Analyst to no automatic safety system function; i.e.,
identify these lhaiting patterns and the it has no scram function. It does provide designated rods either when the patterns the operator with a visual indication of are initially established or as they develop neutron level. The consequences of reactivity due to the occurrence of inoperable control accidents are functions of the initial neutron rods in other than limiting patterns.
flux. The requirement of at least 3 counts Other personnel qualified to perform this per sec assures that any transient, should it function may be designated by the Plant occur,8begins at or above the initial value Superintendent.
of 10~
of rated power used in the analyses of transient cold conditions. One operable C.
Scram Insertion Times SRM channel would be adequate to monitor the approach to criticality using homogeneous The Control Rod System is designed patterns of scattered control rod withdrawal.
to bring the reactor subcritical at A minimum of two operable SRM's are provided a rate fast enough to prevent fuel damage; as an added conservatism, i.e., to prevent the MCPR from becoming
.less than the Safety Limit. The S.
The Rod Block Monitor (RBM) is designed to limiting power transient is that automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.
AmendmentNo.gd,4/,yi,}6,43
... ~
)
3.3 and 4.3 BASES (cont'd) resulting from a turbine stop valve closure later, control rod motion is estimated with failure of the turbine bypass system.
to actually begin. However, 200 mnec 4
Analysis of this transient shows that the is conservatively assumed for this time negative reactivity rates resulting from the interval in the transient analysis and scram (NEDG-24129-1 Figures I and 2) with this is also included in the allowable the average response of all the drives as scram insertion times of Specification given in the above Specification, provide the 3.3.C.
The time to de-energize the required protection, and HCPR remains greater pilot valve scram solenoid is measured i
than the Safety Limit.
during the calibration tests required by Specification 4.1.
The numerical values assigned to the specified scram performance are based on the analysis of The scram times generated at each data from other BWR's with control rod drives refueling outage'and during operation the same as those on JAFMPP.
when compared to scram times generated during pre-operational tests The occurrence of scram times within the limits, demonstrate that the control rod drive I
but significantly longer than the average,
scram function has not deteriorated.
l should be viewed as an indication of a system-In addition, each instant when control l
stic problem with control rod drives especially rods are scram timed during operation if the number of drives exhibiting such scram or reactor trips, individual evaluations l
times exceeds eight, the allowable number of shall be performed to insure that control j
inoperable rods.
rod scram times have not deteriorated.
In the analytical treatment of the transients, D.
Reactivity Anomalies 290 maec are allowed between a neutron sensor reaching the scram point and the start of motion During each fuel cycle, excess operative of the control rods. This is adequate and con-reactivity varies as fuel depletes and as servative when compared to the typical time delay any burnable poison in supplementary control of about 210 maec estimated from the scram test is burned. The magnitude of this excess I
results. Approximately 90 msee of each of these reactivity may be inferred from the critical intervals result f rom the sensor and the circuit rod configuration. As fuel burnup progresses delay, at this point, the pilot scram valve anomalous behavior in the excess reactivity solenoid de-energizer. Approximately 120 msec may be detected by comparison of 103 If, If, f 3[,, f, 43 Amendment No.
i 1
O)
JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that pump shall 2.
Following any period where the LPCI sub-be considered inoperable for systems or core spray subsystems have not purposes satisfying Specifications been required to be operable, the discharge 3.5.A, 3.5.C, and 3.5.E.
piping of the inoperable system shall be vented from the high point prior to the return of the system to service.
4 H. Average Planar Linear llent Generation Rate 3.
Whenever the llPCI, RCIC, or Core Spray (APLMGR)
System is lined up to take suction from the condensate storage tank, the discharge The APLHGR for each type of fuel as a function piping of the HPCI, RCIC, and Core Spray of average planar exposure shall not exceed shall be vented from the high point of the the limiting value shown in Figures 3.5.1 through system, and water flow observed on a 3.5.6.
If anytime during reactor power operation monthly basis.
greater than 25% of rated power it is determined that the limiting value for APLHGR is being ex-4.
The level switches located on the core ceeded, action shal' then be initiated within 15 Spray and RilR System discharge piping minutes to restore operation to within the pre-high points which monitor these lines scribed limits. If the APLHGR is not returned to insure they are full shall be to within the prescribed limits within two (2) functionally. tested every nonth, hours, an orderly reactor power reduction shall be commenced immediately. The reactor power
- 11. Average Planar Linear fleat Generation Rate (APLHGR) shall be reduced to less than 25% of rated power within the next four hours, or until the APLMGR The APLHGR for each type of fuel as a function is returned to within the prescribed limits.
of average planar exposure shall be determined daily during reactor operation at 2: 25% rated thermal power.
4 I
AmendmentNo.g(,]k,43 t
JAFNPP 3.5 (cont'd) 4.5 (cont'd)
- 1. Linear Heat Generation Rate (UlGR) 1.
Linear fleat Generation Rate (UlGR)
'the linear heat generation rate (UlGR) of any The UIGR as a function of core height shall rod in any fuel assembly at any axial location be checked daily during reactor operation shall not exceed the maxisnum allowable UlGR as at > 257, rated thennat power.
calculated by the following equation:
UICR,,x s UERd
( AP/P),,x (L/LT) 1-UlGRd = Design UlGR > G KW/f t.
( AP/P) max = Maximum power spiking penalty = N LT = Total core length = 12 feet l
L = Axial position above bottom of core G = 18.5 KW/ft for 7x7 fuel bundles l
= 13.4 Ky/ft for 8x8 and 8x0R fuel bundles N = 0.026 for 7x7 fuel bundles l
= 0.022 for 8x8 and 8xBR fuel bundles l
If anytime during reactor power operation greater l
than 257. of rated power it is determined that the limiting value for LilGR is being. exceeded, action shall then be initiated within 15 minutes to re-store operation to within the prescribed limits.
If the UlGR is not returned to within the pre-I scribed limits within two (2) hours, an orderly reactor power reduction shall be consnenced launediately. The reactor power shall be reduced to less than 257 of rated power within the next four hours, or until the UlGR is returned to within the prescribed limits.
AmendmentNo.fe, f, 43
JAFNPP O
g i) 8 3.5. BASES (cont'd) 4 requirements for the emergency diesel are within the 10 CFR 50 Appendix K ILmit.
generators.
The limiting value for APLHCR is shown in Figure 3.5.1 through 3.5-6.
G. Maintenance of Filled Discharge Pipe I. Linear Heat Generation Rate (LHCR)
If the discharge piping of the core spray, LPCI, RCIC, and HPCI are not filled, a water hammer can This specification assures that the linear develop in this piping when the pump (s) are started.
heat generation rate in any rod is less l
To minimize damage to the discharge piping and to than the design linear heat generation ensure added margin in the operation of these systems, if fuel pellet densification is postulated.
i
)
this technical specification requires the discharge The power spike penalty specified is based j
lines to be filled whenever the systen is tequired on the analysis presented in Section 3.2.1 to be operable. If a discharge pipe is not filled, of Reference 1 and in References 2 and 3, the pumps that supply that line must be assumed and assumes a linearly increasing variation to be inoperable for technical specification in axial gaps between core bottom and top, purposes. However, if a water hammer were to and assures with a 957. confidence, that no occur, the system would still perform its design more than one fuel rod exceeds the design function.
linear heat generation rate due to power spiking. The LHGR as a function of core H. Average Planar Linear Heat Generation Rate (APLHGR) height shall be checked daily during reactor operation at a25% power to determine if This specification assures that the peak cladding fuel burnup, or control rod movement has temperature following the postulated design basis caused changes in power distribution. For loss-of-coolant accident will not exceed the limit LHGR to be a limiting value below 25% rated specified in 10 CFR 50 Appendix K.
thermal power, the ratio of local LHGR to i
average LHGR would have to be greater than i
The peak cladding temperature following a postu-10 which is precluded by a considerable lated loss-of-coolant accident is primarily a function margin when employing any permissiblu con-j of the average heat generation rate of all the rods trol rod pattern.
of a fuel assenbly at any axial location and is i
only dependent secondarily on the rod to rod power distribution within an assembly. Since expected.
I local variations in power distribution within a l
fuel assembly affect the calculated peak clad 0
temperature by less than 120 F relative to the l
peak temperature for a typical fuel design, the l
limit on the average linear heat generation rate is sufficient to assure that calculated temperatures 1[, ]b, 43 Amendment No.
9 Figure 3.5-5 9
13 3
12 Y6 5w 3$
'z wo 11 0
h 10, 59
~ s l
l 0
5 10 15 20 25 30
(
PIANAR AVERAGE EXPOSURE (GWD/t) l FIGURE 3.5-5 MAXIMUM AVERAGE PIANAR LINEAR HEAT GENERATION RATE (MAPIRGR) VERSUS PIANAR AVERAGE EXPOSURE REFERENCE REIDAD 8DRB265L NEDO-24129 SECTION 14 FULL CORE DRILLED Amendment No. 43 135c l
?
Figure 3.5-6 13 C
O R
12 5
4 ta 35
- z O
11 21 is a gs6 hz 10 Ma 0
5 10 15 20 25 30 t
PIANAR AVERAGE EXPOSURE (GWD/t)
FIGURE 3.5-6 MAXIMUM AVERAGE PIANAR LINEAR REAT GENERATION RATE (MAPIRGR) VERSUS PIANAR AVERAGE EXPOSURE REFERENCE REIDAD 8DRB283 NED0-24129 SECTION 14 FULL CORE DRILLED Amendment No. 43 135D
JAFNPP 4.6 (rgnt'd) 3.6 (cont'd) t
/
)
2.
c.
Frca and aftar tha dzte 2.
at lesat ona cafety/ relief that the safety valve valve shall be disassembled function of one safety /
and inspected once/ operating
^
relief valve is made or cycle.
found to be inoperable, continued operation is 3.
The integrity of the safety /
permissible only during relief valve bellows shall be the succeeding 30 days continuously monitored.
unless such valve is sooner made operable.
a.
The bellows monitoring pressure switches shall b.
From and after the time be removed and hench that the safety valve checked once/ operating function on two safety /
cycle. Modified safety /
relief valves is made or relief valves with two-stage continued reactor operation assemblies do not have a found to be inoperable, bellows arrangement and is permissible only during are, therefore, not subject the succeeding 7 days unless to this requirement.
such valves are sooner made operable.
4.
The Integrity of the nitrogen system and components which 3.
If Specification 3.6 JB.1 and provide manual and ADS actuation 3.63.2 are not met, the reactor of the safety / relief valves shall shall be placed in a cold condition be demonstrated at least once within 24 hr.
every 3 months.
- 4. Low power physics testing and reactor operator training shall be permitted with inoperable components as specified in Item B.2 above, provided that reactor coolant temperature is 5 212 F and the reactor vessel is vented or the reactor vessel head is removed.
r I43 Amendment No. 43
3.6'and 4.6 BASES (cont'd)
JAFNPP f
E.
Safety and Relief / Safety Valves
. i modificd version of the safety / relief valves function with a direct-acting pilot arrangement with no integral Experiences in safety valve bellows.
l operation show that the testing of 50 percent of the safety valves It is realized that there is no way per refueling outage is adequate to repair or replace the bellows to detect f ailures or deterioration.
during operation, and the plant must The tolerance value is specified be shut down to do this. The 30-day in Section III of the ASME Boiler and 7-day periods to do this allow and Pressure Vessel Code as +1 percent of design pressure. An analysis has the operator flexibilit,y to choose his time for shutdown; meanwhile, been performed which shows that with because of the redundancy present in all safety valves set I percent higher, the design and the continuing the reactor coolant pressure safety monitoring of the integrity of the limit of 1,375 psig is not exceeded.
other valves, the overpressure pressure protection has not been The relief / safety valves have two compromised in either case. The f unc t ions; i.e.,
power relief or auto-relief function would not be self-actuated by high pressure.
impaired by a failure of the Power relief is a solenoid actuated bellows. However, the self-actuated function (Automatic Depressurization overpressure safety function would System) in which external instrumenta-be impaired by such a failure.
tion signals of coincident high drywell There is no provision for testing pressure and low-low water level initiate the bellows leakage pressure switch the valves to open. This function is during plant operation. The bellows discussed in Specification B.3.5.D.
In leakage pressure switches will be addition, the valves can be operated removed and bench checked manually.
once/ operating cycle. These bench t
checks provide adequate assurance of The safety function is performed by the same relief / safety valve with bellows integrity. For those modified safety / relief valves with the direct-self-actuated integral bellows and acting pilot arrangement, bellows failures pilot valve causing main valve
~
and bellows related calibrations do not operation. Article 9 of the ASME apply.
pressure Vessel Code Section III -
Nuclear Vessels, requires that these Low power physics testing and reactor bellows he monitored for failure, operator training with inoperable com-since this would defeat the safety ponents will be conducted only when the function of the relief / safety valve.
relief / safety and safety valves are g
Amendment No. 43 152
-s
(
5.0 DESIGN FEATURES B.
The reactor core contains 137 cruciform-shaped control rods 5.1 SITE as described in Section 3.4 of the FSAR.
A.
The James A.
FitzPatrick Nuclear Power Plant is located on the PASNY 5.3 REACTOR PRESSURE VESSEL portion of the Nine Mile Point site, approximately 3,000 ft, east of the The reactor pressure vessel is as i
Nine Mile Point Nuclear Station.
described in Tables 4.2-1 and 4.2-2 The NMP-JAF site is on Lake Ontario of the PSAR, The applicable design in Oswego County, New York, approxi-codes are described in Section 4.2 mately 7 miles northeast of Oswego.
of the FSAR.
The plant is located at coordinates north'4,819,545.012 m, east 386,968.945 m, 5.4 CONTAINMENT on the Universal Transverse Mercator System.
A.
The principal design parameters and characteristics for the B.
The nearest point on the property primary containment are given in line from the reactor building and Table 5.2-1 of the FSAR.
any points of potential gaseous effluents, with the exception of the B.
The secondary containment is as lake shoreline, is located at the described in Section 5.3 and the northeast corner of the property.
applicable codes are as described This distance is approximately in Section 12.4 of the FSAR.
3,200 ft. and is the radius of the exclusion area ar defined in 10 CPR C.
Penetrations to the primary con-100.3.
tainment and piping passing through such penetrations are designed in 5.2 REACTOR accordance with standards set forth in Section 5.2 of the PSAR.
A.
The reactor core consists of not more than 560 fuel assemblies.
For 5.5 FUEL STORAGE the current cycle three fuel types are present in the core:
- 7x7, A.
The new fuel storage facility is 8 x 8, and 8 x 8R.
These fuel types designed so that the Keff dry is are described in Section 3.2 of the
< 0. 90 and flooded is < 0.95 des-FSAR, NEDO-20360, and NEDO-24011, cribed in Section 9.2 of the PSAR.
respectively.
The 7 x 7 fuel has 49 fuel rods, the 8 x 8 fuel has 63 fuel rods and I water rod, and the 8 x 8R fuel has 62 fuel rods and 2 water rods.
Amendment No. )MI, 43 245
.